This invention relates to the pretreatment of a radioactive liquid waste produced by reprocessing or generated by a light-water or breeder reactor, wherein the liquid waste contains mainly sodium compounds and is disposed in the form of a solid in a mixture with glass, asphalt, plastic or the like. More particularly, the invention relates to a process for electrolytically treating such a radioactive liquid waste for the purpose of recovering useful elements, such as sodium.
The high-level liquid waste produced by reprocessing or generated by a light-water or fast breeder reactor is a solution comprising 40% Na.sub.2 O and 60% fissile material, actinides, corrosion products and the like. Conventionally, a glass-forming material is added to such high-level liquid waste and the mixture is subjected to a heating process to solidify the glass. Alternatively, the liquid waste is mixed with an asphalt or plastic material and solidified therewith and the resulting solid is discarded at a storage location. For examples of such prior art, see the specifications of Japanese Patent Publication (KOKOKU) Nos. 39-15100, 42-15519. In order to remove sodium compounds from these liquid wastes, processes have been proposed in which the high-level liquid wastes are heated to remove the sodium compounds by decomposition and vaporization.
However, it has been pointed out that the solidification processes using glass or the like are disadvantageous in that there is an undesirable increase in bulk owing to addition of the glass, etc. forming materials, and in that it is difficult to recover useful nucleides from the glass solid, or the like. It has also been pointed out that the process for decomposing, vaporizing and removing the sodium compounds by heating the high-level radioactive liquid waste also involves such problems as the need for high temperatures, the decomposition and vaporization of nucleides having low boiling points, and the complexity of the off-gas treatment system.